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Activities in the NUCLEAR AREA have been focused on the realization of thermo-hydraulic studies and safety analysis for the different types of nuclear power plants. During these years we have developed and updated different simulation software and have made several experimental works.


Thermalhydraulic simulations of accidents and operational transients in conventional nuclear power plants and new plant designs are performed using the most current versions of the codes RELAP5, TRACE and TRAC-BF1.

We participated in projects for the European Commission (EC), IBERDROLA, Nuclear Safety Council of Spain (CSN) UNESA, CIEMAT, Central Nuclear de Trillo Nuclear Power Plant, etc. Major researches in this line are:

1. Study of the codes capability to reproduce complex experiments. For some time now we are cooperating with the NRC (Nuclear Regulatory Commission) in the United States in the CAMP project (Code Applications and Maintenance Program). To verify that the codes correctly reproduce the phenomena several experiments are performed in large facilities and then run with the codes. Recent experiments, performed in the PKL and ROSA facilities, have been modeled and the results are compared with their experimental data.

2. Study of averaging techniques for thermalhydraulic parameters in mixed cores. We continued with a study for the Spanish Nuclear Safety Council on this issue. In a previous study we had determined how the average of the thermalhydraulic magnitudes of different types of nuclear fuel should be done, and during these three years we study the effect that this new averaging method has on the value of the CPR (Critical Power Ratio) when we average the gap heat transfer coefficient (HGAP) of a core with a large number of fuel types.

3 Scale analysis in thermal hydraulics installations. To obtain the uncertainty in the calculations we need to compare with experiments that have been conducted at a scale smaller than the real (lower than real dimensions). It is therefore necessary to make a scale analysis. We developed a complete scaling methodology, and it has been applied to the case of long-term cooling accident LOCA (loss of coolant) in a passive BWR reactor.

4. Animated Interactive Model of Cofrentes NPP. Using new tools currently in development (TRACE-SNAP), a thermalhydraulic Simulator of Cofrentes Nuclear Power Plant for the IBERDROLA Group was created. The model created, which it is continuously updated and improved, allows the simulation of operational transients such as: the runback of the feedwater turbo-pumps, with simultaneous closure of MSIVs with close failure of one SRV, or runback both recirculation pumps.

5. Modeling and analysis of transients of Trillo NPP. During these three years we continued improving and updating the Trillo NPP model made in TRACE, and several transients were simulated, as the pressurizer pre-operational tests and the scram transient do to the transfer of 400 KV to 132 KV (RESA).

6. Thermalhydraulics of severe accidents. Within this area we have studied the hydrodynamics of gas discharges in aqueous beds (jets) to see how to calculate the amount of radioactive particles that are captured by water. In a severe accident this discharge occurs, for example, when a steam generator tube ruptures. We have studied the codes that are currently in use and that we have updated and improved. This work was carried out within the international project ARTIST-II in collaboration with CIEMAT.


Reactor stability studies are considered in this line of investigation under three different approaches:

The first way is using computer codes to simulate the behaviour of the reactor core, such as LAPUR and PAPU (both licensed by the Nuclear Safety Council of Spain). These programs have been developed, partial or totally, by the members of this group of investigation, and are being used by the nuclear industry (IBERDROLA). With these tools it is possible to simulate the reactor behaviour, both, for real and fictitious conditions. A second approach is based on the calculation of the decay ratio -DR- from neutron detector signals of the plant by using autoregressive methods. This technique allows a complete and real knowledge of the plant (a posteriori). Finally, the third form is by the use of the DWOS code (developed for this IIE group) that permits the no-lineal analysis of the nuclear reactor behaviour.

In this line, it is worth mentioning the work that is being developed for the IBERDROLA group, in the project DROP. This project consists of the development of an on-line predictor for the stability of the Cofrentes Nuclear Power Plant. The basic element of this predictor will be an improved version of the LAPUR code. Among the improvements that have been studied we can highlight: the calculation of the void fraction using new correlations (i.e. Zolotar Lelouche); implementing the dependence of the conductivity of uranium with burnup. In addition, a new three-dimensional kinetic model to LAPUR is under development and implementation. 

Also note that we have continued to improve and update the SMART monitor, developed for IBERDROLA, which is capable of analyzing several plant signals coming from the APRM and LPRM detectors, and calculate the reactor stability from the signals. To verify and validate the monitor a program (GeneradorBWR), that generates synthetic signals using a reduced order model, has been developed.


Simulations of the behavior of single-phase and multiphase fluids inside complex structures are addressed in this line of activity, using fundamentally the CFX commercial code. During this period two works should be highlighted out:

The research group has participated in two OECD benchmarks.

The first consisted of a T-junction, in which for one branch fluid circulates at a given temperature and on the other side branch the fluid is injected at a higher temperature; the goal is to predict the mixing, the velocity distribution downstream of the main branch and the temperature fluctuations over time in different sections, as well as the power spectral density of these fluctuations. The results were very good.

The second Benchmark consisted in predicting the velocity distribution and the velocity fluctuation produced by two kinds of spacers of a PWR fuel assembly. Very good results were also obtained despite the complexity of the cases analyzed.


In this line of investigation Monte-Carlo techniques are applied to obtain the multiplication factor of nuclear systems. The MCNP codes and the SCALE system are used to determine the constant of multiplication in nuclear systems, such as new and spent fuel storage pools, accelerator driven subcritical system, etc.

In this line we can highlight the work done within the CDT project (Central Design Team) of the 7th framework of the EEC for the development of an incinerator for radioactive waste cooled with Lead-Bismuth in eutectic state (FASTEF-Fast-spectrum Transmutation Experimental Facility); which can operate in critical mode, as fast reactor, and subcritical mode, powered by a source of neutrons produced by spallation. In collaboration with KIT and ENEA various transient that may occur in the FASTET normal operation mode have been simulated.


IIE has the expertise to provide support for the licensing of nuclear fuel reload in topics such as LOCAs, ATWS, etc. There are also experience and codes to carry out the analysis of radiological consequences in base design accidents and severe accidents, both for LOCAs and SGTR. The NUREG guide-1465 and the EUR normative are followed in all these analysis. 

In this line of research it has also been conducted for the Spain Nuclear Safety Council, a study of the methodology CSAU (Code Scaling Applicability and Uncertainty) to assess its applicability using Best Estimate codes in licensing activities. Basically, the CSAU methodology consists of the evaluation of the capacity of a specific code in the simulation of a particular scenario in a plant. In particular applied to an ATWS transient (Anticipated Transient Without Scram) of a BWR plant. To apply this methodology, the sources of uncertainty that might exist in the plant model, physical properties, close relationships, physical models, numerical methods and nodalization were studied. Based on this study and the PIRT (Phenomena Identification and Ranking Table) a first selection of the most important phenomena was selected, and their corresponding code parameters. Once these parameters are known, two codes (GEDIPA and UNTHERCO) are developed, and allow us to obtain, from all statistical information available for that parameter, a set of values for these parameters to be used in the Best Estimate code, and obtain the uncertainty in the calculations performed.


There is experience and capacity for analysis of new nuclear reactors. The work done in these years include those made for the following reactors:

Lead-Bismuth Cooled Reactors. In this line the Lead-Bismuth thermalhydraulic has been studied. A number of operational transients were realized for the project of XT-ADS that consisted of 3 loops, one first for the lead-bismuth transferring heat to a secondary water forced convection system, and from this secondary loop the heat is transferred to a air-cooled loop.

Sodium cooled reactors. The capabilities of RELAP5-Na and Astec-Na for simulating severe accidents of this type of reactors are being studied. To do this many transients of the CABRI facility and the PHENIX reactor are modeled and run with the code RELAP5-Na, and to see if the thermalhydraulic models of this code allow the program to correctly reproduce the phenomenology that occurs in these cases. This work was carried out within the international project Jasmin in collaboration with CIEMAT.


In this line we can highlight different facilities that have been made. All of them, depending on the experiment that was intended to study, are instrumented with pressure and temperature sensors, conductivity sensors, flowmeters, etc..; and the imaging systems (high-speed cameras, LASER Anemometry, stroboscopic light sources) suitable to capture with great detail the phenomenology have been used. Within the facilities developed the most important are:

Facility to study the safety discharge in a pool. In this installation behavior of a jet produced by the discharge of air through nozzles of different diameters and with different flow rates of air was studied. We intended to reproduce, scaled, the behavior of the suppression pool of the nuclear power plants. This facility has been financed by the National R & D Plan 2011-2013, within the project REMODERN.

Installation for the study of air entrainment by a vertical jet of water discharge in a pool. This facility is used to characterize the amount of air that is entrained by the jet and its behavior in the pool. This facility has been financed by the National R & D Plan 2011-2013 within the project REMODERN.

Installation for the study of countercurrent flow. This is the latest installation, and it is intended to study the behavior of a film of water with a countercurrent flow of air, as in the case of stratified flow during a small LOCA.